ML20207G590

Forwards Response to Request for Addl Info Re Plant Safety Relief Valve Discharge Test Results Per Testing Procedure Recommended in NUREG-0661 & NUREG-0763

ML20207G590
Person / Time
Site:	Fermi
Issue date:	08/12/1988
From:	Sylvia B DETROIT EDISON CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0661, RTR-NUREG-0763, RTR-NUREG-661, RTR-NUREG-763 NUDOCS 8808240169
Download: ML20207G590 (4)

Text

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8. Ralph Sylvia

, Senor Vice Presdent DOITOi,I . ~,n o,, ,, % n . .,

Edison ~==~~ x -

X, 4 August. 12, 1988 NRC-88-0194 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) NRC Letter to Detroit Edison, dated July 13, 1988

3) Detroit Edison Letter to NRC, "Submittal of-Safety / Relief Valve In-Plant Testing Results," ,

NRC-87-0137 dated September 11, 1987

4) Detroit Ecison Letter to NRC, "Submittal of SRV In-Plant Test Plan." EF2-59,029, dated August 18, 1982

Subject:

Response to Request for Additional Information Regarding Fermi 2 SRV Discharge Test Results In the Reference 2 lotter the NRC transmitted a list of questions regarding Fermi 2 Safety Relief Valve tests and asked Detroit. Edison to provide a response. Enclosed please find our response, i

If you have any questions regarding this response, please contact Mr.

Lewis Bregni at (313) 586-4072.

l Sincerely, Enclosure cc: Mr. A. B. Davis Mr. R. C. Knop Mr. T. R. Quay Mr. W. G. Rogers

! I 8808240169 880812 PDR ADOCK 05000341 ,

p PDC ,

1 1

Enclosure to NRC-88-0194 Page 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING FERMI 2 SRV DISCHARGE TEST RESULTS NRC Question No. 1 Testing Procedure for Fermi 2 compared to that recommended in NUREG-0661 and NUREG-0763:

The above NUREG's have provided guidelines for plant-specific tests (e.g. Section 6 of NUREG-0763, "Test Procedures and Matrix"). It is not clear from the test report whether the testing sequence and conditions used for Fermi 2 satisfy the NUREG recommendations (e.g.

discharge conditions, number or quenchers, thermal mixing, simultaneous actuation of multiple valves).

Detroit Edison Response The Fermi 2. test. plan was submitted to the NRC by letter dated August 18, 1982 (Reference 4). A copy.is enclosed. As noted in the test plan, the scope of the-SRV discharge test for Fermi 2 was limited to confirming that the methodology used in the Fermi 2 Plant Unique Analysis Report (PUAR) for evaluating SRV discharge related loads was conservative. The test scope was limited because the methodology in the PUAR was developed using the alternate criteria for SRV discharge loads contained in NUREG-0661, and was based upon conservative analysis methods previously confirmed from other in-plant tests. As a result, the criterion and guidelines contained in NUREG-0763 for performing limited confirmatory in-plant tests are applicable.

The significant reason for performing plant specific testing for Fermi 2 is found in NUREG-0763, Part 4 "RATIONALE FOR PLANT SPECIFIC TESTS", Item (1): "The discharge device is geometrically different from devices tested previously." The remaining reasons the NUREG tists for performing plant specific tests were able to be treated

• nerically at Fermi 2. The enclosed test plan discusses how the test performed at Fermi 2 satisfies NUREG-0661 and NUREG-0763 NRC Question No. 2 Testing Temperature:

NUREG-0763 recommends discharge under the normal temperature condition. It is not clear whether this condition has been satisfied for Fermi 2 (Rof. NUTECH report p4-1).

. x Enclosuro to' NRC-88-0194 Page 2 Detroit Edison Response NUREG-0763, Part 6.0 "TEST PROCEDURES AND MATRIX", recommends that plant-specific tests focus on single valve actuations under normal discharge conditions (cold pipe, normal water leg). It should be noted that during Fermi specific testing the temperature readings at temperature sensors T5 and T6 located on the SRV discharge line within the torus air space were indicating torus ambient temperature.

Additionally, an air bleed system was installed at Fermi 2 and was used during testing to adjust the water level before each discharge.

Part 6 of NUREG-0763 also recommends that plant specific tests generally not include leaking valve actuations: referred to in the NUREG as LVAs. The NUREG further recommends that a temperature sensor near the SRV should be monitored to avoid testing under LVA conditions and that load changes associated with these conditions should be quantified on a generic basis. Page 4-1 of the test raport. (Reference

3) discusses the test results and notes that all but one of the single valve actuation (SVA) discharge tests was performed with an SRV tailpipe temperature of approximately 212 F, which is indicative of a leaking valve. However, it was judged in the. test report that the leak was small enough that the air bleed system would equalize the water leg and the pressure buildup was slow enough that test results would not be significantly affected.

The one exception was test MT1, which was performed at a ta!1 pipe temperature of 167UF. Recordings taken during normal plant operation indicate that 167 0F is within the normal ambient range of SRV tailpipe temperatures. . Thus, test MT1 was not a leaking valve actuation (LVA). Section 4.1 of the test report states the only observed difference in the results of test MT1 and the other SVA tests is an upward shift of the dominant bubble frequency, which ranged from 7 to 11 Hz, for the SVA tests with elevated SRV tailpipe temperatures. This shift in dominant bubble frequency is toward the fundamental frequency of the suppression chamber and results in a conservative shift in the discharge loads. The elevated tailpipe temperature conditions therefore had the positive consequence of providing more conservative data for comparison to the analytical values.

)

l

Enclosura'to .

l NRC-88-0194' Page 3 NRC Question No. 3 90-90 vs. 95-95 statistical Results:

The test data in the report were analyzed for a 90-90 probability value rather than the customary 95-95 estimate. Will the 95-95 results have any impact on the conclusion?

Detroit Edison Response The methodology used to determine SRV discharge loads for Fermi 2 was based in part on results obtained from Monticello in-plant tests. A 90-90 statistinal analysis was performed on the Fermi 2 test data because that is the same analysis used for Monticello.

Although a 95-95 statistical analysis would bound more data and consequently have a higher variance with reduced margins, the quality of the data is such that the magnitude of the change would be small.

.Therefore, a 95-95 -statistical. analysis of the test data.would have no significant impact aon the existing . conclusions.

NRC Question No. 4 Analytical Results used for comparison with the Test Data:

Provide a summary comparison of the test results and the original analysis results for several of the most highly stressed points in the structure.

Detroit Edison Response Section 5.0 "RECONCILIATION OF PUA RESULTS" of the test report (Reference 3) discusses the test results as compared to the i analytically determined loads. Table 5-2 of the same section provides l a comparison of the maximum measured stresses with analytically predicted results for each group of the strain gauges, i 1

8." '

Harry T:ube, tw .w~~

Deyolt

• ECiSOn inUBN" -

August 18, 1982 EF2 - 59,029 , *

~

Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing.

U. S. Nuclear Regulatory Commission Washington, D. C. 20555

~

References:

(1) Enrico Fermi Atomic Power Plan" % Unit 2 NRC Docket No. 50-341 l l

(2) NRC letter, May 26, 1982, B. J. ,

Youngblood to H. Tauber, "Mark I i Containment Analysis

• I l

Subject:

Submittal of SRV In-Plant Test Plan I

)

At your request (Reference 2), we are submitting the attached Fermi 2 SRV In-Plant Test Plan.

l There are only approximately twenty weeks remaining I before the final scheduled torus fill before startup.

As such, if your staff has any comments relative to our proposed test which may affect the test instru-mentation in the torus, we need to resolve them as expeditiously as possible in order not to impact the Fermi 2 startup schedule.

would thus be appreciated. Your immediate attegtton- l l

If you have any questions regarding the above, please  !

contact L. E. Schuerman, (313) 649-7362.-

l Sincerely, Attachment cc: Mr. L. L. Kintner Mr. B. Little on .no__ //

C74(./O KV{N > kf

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j Mr. B. J. Youngblood i 59,0291982 18 l Auhust-EF i Page 2 bec: T. A. Alessi E. L. Alexanderson H. O. Arora M. L. Batch l e, s J. H. Casiglia L. E. Eix l W. J. Fahrner E. P. Griffing W. R.' Holland W. H. Jens  ;

L. E. Kanous l J. Levine  ;

A. K. Lim ,

E. Lusis  !

P. A. Marquardt i J. W. Nunley '

L. E. Schuerman l R. A. Vance A. E. Wegele O. K. Earle Document Control F. E. Gregor (MDC-Southfield) .

C. M. Johnson (GE-San Jose)

D. F. Lehnert (MDC-Southfield)

R. Roberts (NUTECH-San Jose)

J. E. Slider (NUS-Gaithsburg)

F. H. Sondgeroth (NUS-Troy) l l

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DET-22-014 Rnvision 0 July, 1982 Safety Relief Valve In-Plant Test Plan -

For the Enrico Fermi-2 Atomic Power Plant ,

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Unit 2 s

Prepared for:

Detroit Edison Company Prepared by:

NUTECH Engineers,.Inc.

San Jose, California Issued by: Issued by: ,

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'b J . R. Leonard, P.E. C. W. Roberts i Project Engineer Project Director 1 i

Approved by:

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R. A. Lehnert, P.E.

Engineering Manager p g) nnf}nn1

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Fermi-2 SRV In-Plant Test Plan ,

. Table of Contents

. Page 1.0 Int'roduction 1 2.0 Test Objectives and Sedpe 1 44 3.0 Test Program Rationale and Instrumentation 4 3.1 Plant Unique Analysis Report (PUA) Model 4 3.2 Test Bay Selection 6 3.3 Test Instrumentation 7 4.0 Test Program 10 4.1 Procedures 10 4.2 Test Matrix 10 4.3 Data Reduction and Reports 12 5.0 Program Schedule 13 6.0 Re ferences 13 4

DET-22-014 i Revision 0 .

. l Fermi 2 SRV In-Plant Tant Plan

'l 1.0 Introduction

. Fermi 2 is equipped with safety relief valves 45RV's) to

. control primary system pressure transients. Wh'en a SRV is actuated, steam is released from the primary system

~

and compresses the air within the safety relief valve discharge line (SRVDL) . This compressed air enters the pool in the form of high pressure bubbles which oscillate, resulting in pressure loads on the torus shell and internal structures. Section 1-1.3.2 of the Fermi 2 Plant Unique Analysis Report (PUAR) (Reference 1) provides a more detailed ,

discussion of the SRV discharge phenomena. The evaluation of the Fermi 2 SRV discharge loads presented in the PUAR i

utilized the alternate methodology contained in Sectiou

]

2.13.9 of Appendix A of NUREG-0661 (Reference 2) . As such, SRV in-plant tests will be performed for Fermi 2 after fuel  !

load to confirm that the methodology used in the PUAR for evaluating SRV discharges is conservative. This document describes the planned SRV in-plant tests for Fermi 2.

I 2.0 Test objectives and Scope The objective of the Fermi 2 in-plant SRV discharge test is to confirm that the loads and structural responses documented in the Fermi 2 PUAR for SRV discharge related loads are conservative compared to the loadings and structural responses which occur during actual SRV discharges.

DET-22-014 Revision 0 , 1 i

Tha test progrcm planncd for Fsrmi 2 10 boing devalopsd in accordance with NRC guidelines for in-plant tests ,

contained in NUREG-0661 and NUREG-0763 (Reference 2 and 3).

A discussion of the applicability of these guidelines to Fermi 2 is contained in the paragraphs which' follow. -

~

The characteristics of the T-quencher device being used in Fermi 2 are the same as those used in other Mark I plants and tested at Monticello. The SRV discharge methodology contained in NUREG-0661 and that which is based on results obtained from Monticello in-plant tests are applicable for use at Fermi 2, as discussed in the PUAR. The scope of the Fermi 2 in-plant test is limited to confirmation of the SRV discharge methodology used in the Fermi PUAR. This method-ology was developed using the alternate criteria for SRV discharge loads contained in NUREG-0661. The criteria and guidelines contained.in NUREG-0763 for performing limited confirmatory in-plant tests are therefore, applicable.

Conservative analysis techniques are utilized to demonstrate acceptable Fermi 2 torus attached piping and submerged structures response to SRV loads. Section 2.13.9 of NUREG-0661 states that test measurements are not needed if such conservative analysis techniques are used. Therefore, g the Fermi 2 SRV in-plant test does not include measurement of torus attached piping or submerged structures response to SRV loads.

DET-22-014 Revision 0 2

. - - . - - - . . _ . . __ . _. x . .-_ -.

The SRV in-plant test plannsd for Farmi 2 will focus on l the following areas: i s

1) Measurement of the line clearing reaction loads l on the SRV discharge line and T-quencher supports. #

2) Measurement of peak pool boundary pressures during air clearing and steam discharge due to a single t valve discharge -(normal water leg, cold pipe) s,,

3) Measurement of the frequency content of the T-quencher air-bubble-transient pressure signatures. l The SRV discharge nethodology used in the Fermi 2 PUAR utilizes generically developed conservative methods as defined by NUREG-0661 for determining spatial variations in air-bubble loads, load superposition methods for evaluating multiple valve actuations, load changes that accompany consecutive valve actuations, and shifts in I

bubble frequencies that result from variations in back pressure during air clearing. Therefore these methods do not require confirmation by in-plant test. Supp-ression pool thermal mixing tests will not be conducted since the evaluation of pool temperature response to SRV )

transients described in the PUAR demonstrates compliance with the required pool temperature limits. Additional specific items included in the Fermi 2 in-plant test are discussed in the sections which follow. 4 DET-22-014 3 l Revision 0 l l

= _ _ _ _ _ - . _ _ . _ . _ . _ _ _._.___ ._.. .

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' 3.0 Tant Program Rationala and Instrumantation The test procedures and instrumentation planned for the -

Fermi 2 SRV in-plant test are being developed in accordance with the NRC guidelines for in-plant tests contained in NUREG-0661 and NUREG-0763 (References 2 and 37. Since'a "l'arge data base has already been' developed for evaluating the effects of SRV discharge in a Mark I containment, the

• u scope of the instrumentation end testing for Fermi 2 will ,

be limited to that necessary to meet the confirmation objectives discussed in Section 2.0. The key elements of the Fermi in-plant test program are discussed in the sections which follow.

3.1 Plant Unique Analysis (PUA) Model The analytical approach used to evaluate the response 1 <

of the suppression chamber to SRV discharge torus shell loads is discussed in the Fermi 2 PUAR. The approach utilizes a coupled lead-e structure analytical model developed.in accordance ,

with NUREG-0661 criteria. -

Predicted SRV discharge torus shell pressure load magnitudes and time characteristics are developed using an analytical model based on Monticello in-plant test data.

i It was demonstrated in the l PUAR that the analytical model results in load I

magnitudes which envelop those measured in the Monticello tests.

, DET-22-014 4

Revision 0

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i The structural evaluation of the suppression  !

chamber for SRV discharge. torus shell loads is performed using a finite element model of i i

a representative segment of the suppression

• l

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chamber. The analytical model includes a finite  !

element representation of the suppresison pool to account for fluid-structure interaction effects.

A forced vibration analysis of the suppression chamber is performed for the Monticello-based SRV torus shell loads using the finite element model discussed above. Calibration factors, developed using Mon?.icello in-plant test results are applied to the results to convert the forced vibration response to a free vibration response.

The approach is based on the observation that the -

phenomena associated with an SRV discharge into the suppression pool are characteristic of an initial value or free vibration rather than a forced vibration condition. The calibration factors used in the Fermi 2 suppression chamber analysis for SRV discharge torus shell loads are documented l

in the PUAR and will be confirmed using Fermi 2' i I

in-plant test results. l DET-22-014 5

, Revision 0 4

4

1 3.2 Test Bay Selection 9

The SRV discharge line (SRVDL) se'lected for the in-plant test is one of the shortest in length, measured from 1

the SRV to the T-quencher ramshead along the, pipe center-line. Compared with other SRVDL's the line contains the smallest volume of air between the SRV and the submerged portion of the SRV piping in the wetwell.

1 This characteristic results in air clearing loads which are closer in frequency to the dominant frequency of the suppression chamber than those produced by longer SRVDL's.

1 As the frequency of the SRV discharge torus shell loads approach the dominant frequency of the suppression chamber, the dynamic response of the suppression chamber increases.

Although the frequencies of the air clearing los ds are expected to be less than the dominant frequency of the suppression chamber, use of the selected SRVDL is expected to maximize dynamic amplification effects and provide a basis for confirming the model calibration i factors discussed in Section 3.1.

The magnitude of the SRV discharge air clearing loads "

used in the Fermi 2 PUAR are developed using Monticello-based analytical models, as discussed in Section 3.1.

The analytical models account for different line geometries and other parameters which affect load DET-22-014 6 Revision 0

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characteristics'. The SRVDL selected for the Fermi 2 in-plant test will provide sufficient basis to confirm that the load magnitudes used in the PUAR*are conservative and, in general, confirm the adequacy of the analytical

)

approach used in the POAR to develop SRV discharge air clearing loads.

The suppression chamber segment selected to be instrumented is located away from major suppression chamber attachments which may have an affect on t'he suppression chamber frequency. The structures contained in the selected suppression chamber segment are representative of those contained in al'1 suppression chamber segments.

3.3 Test Instrumentation In order to meet the test objectives, measurements must be made of SRV discharge bubble pressures, T-quencher internal pressure, torus shell pressure and response of the torus shell and support system.

To accomplish this, pressure transducers will be placed on the T-quencher arms to measure it?v~7al and external source pressure, on the torus shell to measure the resulting torus shell pressure, and inside the SRVDL to ensure that the SRVDL water level for subsequent actuations has returned to a steady state value following SRV closure. Transient reflood l

l DET-22-014 .

7 l Revision 0 ,

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characteristic ~c"will'n'o't bo maasurGd cinco o normal ~

SRVDL water lovel io casumsd in the PUAR for all

. . subsequent actuation cases. Strain gages will be -

,. placed on the internal and external surfaces of k the torus shell in a rosette configuration to provide measurements of extreme fiber and 'embrane m

stress intensities on the shell. The pressure 5,, tranducers and strain rosettes located on the torus shell are fewer in number but arranged in a manner similar to those of the Monticello in-plant test.

Uniaxial strain gages will be located on the columns and saddle supports in the test bay to -

record the total integrated reaction load of the suppression chamber support system. Uniaxial strain gages will also be located on the support columns in the adjacent bays to measure attenuation b effects. A summary of sensor characteris. tics is provided in Table 3,1-1. An air bleed system will be installed on the SRVDL in the drywell to equalise the pressure between the discharge line and the drywell air space prior to SRV actuation. The existing plant SRVDL temperature and pressure sensors will be used to detect a leaking SRV. C 4

The test instrumentation will require approximately 125 recording channels with a maximum frequency response of 200 hz. The signals from the instru-('- DET-22-014 8 Revision O ,

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Table 3.3-1 -

Summary of Sensor Characteristics Sensor Type Location Range Environment A. Pressure transducer ,

1) low pressure Torus shell, external 0-100 psi Watgr, air, steam 9 50 psi &

quencher, air bleed 270 F (max)

,, 2) Iligh pressure Internal quencher, 0-1000 psi Water, air, steam @ 700 psi &

SRVDL 400 F B. Strain Gages -

1) Weldable Internal torus shell, 0-0.02 in/lii Watgr, air, steam 6 50 psi &

quencher supports 270 F (max)

2) Foil External torus shell 0-0.D2 in/in Air @ 14.7 psi & 100 F (max) e

. e 9

b - . , _. _ - . _ _ _ -

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mentation will ba procascod by approprinto cignal

, conditioning equipment and stored on magnetic

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. tape in digital format. Each sensor will be scanned at approximately 1000 samples per second..

. Approximately 25 percent of the channels sill be 1

processed on site for comparison with the test I

, acceptance criteria.

4.0 Test Program l l

4.1 Procedures l 1

Necessary procedures will be provided to outline l the requirements for sensor placement and installation, qualification of test personnel, calibration of I instrumentation, establishing pre-test conditions and conduct of the matrix tests.

4.2 Test Matrix The test matrix is presented in Table 4.2-1.

Shakedown test (s) will be conducted to verify operational procedures, to optimize test operations and to establish recorder settings for realtime instrumentation. Matrix testing consists of at least four test pairs to evaluate the effects cf a single valve actuation (SVA) and a subsequent c consecutive actuation (CVA) of the same valve. The number of SVA/CVA test pairs to be conducted will depend upon the data scatter encountered as determined by a statistical review of the real-time data DET-22-014 10 Revision 0

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Table 4.2-1 .

"[ .

gg TEST MATRIX o tn ,

<q IN1TIAL, CONDITIONS L

F 8 VALVE P1PE VALVE -

CIASURE rU O L TEST NUMUER TEST TYPE TO EE ACTUATED SRV PIPE POOL POWER DISCHARGE TIME Pet 10R COOLisco PRIOR TO U O TEMP (

• F) LEVEL (1) TING (SEC) TO CVA TEST o SD1 SD SRV-2066 CP, NWI, See Note 1 See Note 2 10 N/A *See Note 3 Mit SVA CP.NWL 11 N/A See Note 3 j MT2 CVA HP, AWL 10 1 Min 1 Min MT) SVA CP.NWL 10 N/A See Note 3 MT4 CVA HP,AW1. 10 1 Min 1 Min d

Mr5 SVA CP,NWL 10 N/A See Note 3 MT6 CVA HP, AWL 10 1 Min 1 Min wr7(4) SVA CP.NWL 10 N/A See Note 3 MTS(5) CVA HP,AWt, 10 1 Min 1 Min MT9 SVA CP,NWL 10 N/A See Note 3 MTIO CVA HP, AWL 10 1 Min . 1 Min i H l Notees (1) The start 1ng pool temperature le unknown prior to the test but all teste shall be run with pool temperature i 10* F of the starting temperature (2) Power levet sufficient to support steady steam flow through an SRV discharge line at 1000 pet at the SRV. -

i f,3 ) Pipe temperature to be within i 10* F of the temperature before test NFL or at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of cooling should be provided following a previous actuation. .

(4) Af ter performing test Mr7, additional SVA teste shall not be performed unless required.

(5) After performing test MS, additional CVA teste shall not be performed untees required.

Abbreviatione: SD - Shake Down SVA - Single Valve Actuation CVA - Consecutive Valve Actuation CP - Cold Pipe (SRVOL)

IIP - Hot Pipe ( S RVDI.)

NWL - Normal Water Level (In the SRVDL) ,

AWL - Actual Water Level (In the SRVDL) .

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tollowing tha fourth'and uubscqusnt tocto but will not GxcGCd cix tssto. Rsal tim 3 data will also be evaluated against acceptance criteria

• 1 after each shakedown test and during the cooldown period following each pair of SVA/CVA tests.

. The duration of discharge selected for eaah test will be 10 seconds with 60 seconds between SVA and CVA tests based upon an analysis of the trends reported in Reference 3.

4.3 Data Reduction and Reports In addition to insitu data processing, filtered

' ~

time history and power spectral density plots

- -in engineering units will be generated for each data channel and for all tests. Further post processing to produce frequency averaging over

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several tests, component bending and axial strains for a particular gage location, and stress intensities for strain rosette configurations will be conducted as required.

A final report will be prepared which will contain (1) a discussion of the instrumentation locations, calibrations, signal conditioning system, instrument uncertainty, data collection and reduction, (2) a tabulation of maximum and minimum values of ,

all data channels for each test condition,'

(3) A discussion of test results and comparison to l

(

expected results, (4) representative plots of all DET-22-014 12 Revision 0

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, dato in engincaring unito for all conditiono

' . .e tostod, cnd (5) a summary of conclucions regarding the confirmation of the Fermi PUA for SRV discharge loads.

S.O . Program Schedule ~

, Testing will be conducted with the' plant operating on the bypass system at 10% power or_ greater and will span a period of from one to two weeks (actual test time will not exceed three days including shakedown tests).

Data reduction and analysis will follow and culm'inate in the issuance of the final report.

6.0 References

1. NUTECH report DET-04-028-1, "Znrico Fermi Atomic Power Plant Unit 2 Plant Unique Analysis Report, Volume 1", Revision 0, April l'982.

2. NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program", July 1980.

3. NUREG-0763, "Guidelines for Confirmatory Inplant Tests of Safety-Relief Valve Discharges for o

BWR Plants", May 1981. -

DET-22-014 13 ,

Revision 0 1

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